Text

Amend 61 to License NPF-42,revising TS to Support Use of Vantage 5H Fuel W/Intermediate Flow Mixers
	ML20035C619
Person / Time
Site:	Wolf Creek
Issue date:	03/30/1993
From:	Black S; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20035C620	List: ML20035C619; ML20035C621;
References
	NUDOCS 9304080196
	Download: ML20035C619 (92)
	v • d • e

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jg UNITED STATES t

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NUCLEAR REGULATORY COMMISSION

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.E WASHINGTON, D. C. 20555 t

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WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50 482 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 61 License No. NPF-42 1.

The Nuclear Regulatory Commission (the Comission) has found that:

A.

The application for amendment to the Wolf Creek Generating Station (the facility) Facility Operating License No. HPF-42 filed by the Wolf Creek Nuclear Operating Corporation (the Corporation), dated October 28, 1992 and suplemented by letters dated January 28, 1993 and March 8,1993, comp les with the standards and requirements of the Atomic Energy Act of 19 i, as amended (the Act), and the Comission's rules and regulations se, forth in 10 CFR Chapter I; B.

The facility will operate.3 conformity with the application, as amended, the provisions of s he Act, and the rules and regulations of the Comission; C.

There is reasonable assuranct*

(1) that the activities authorized by this amendment can be conduct d without endangering the health and safety of the public, and (ii) + hat such activities will be conducted in compliance with the Comission > regulations; D.

The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

9304080196 930330 PDR ADDCK 05000482 P

PDR

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2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Facility Operating License No. NPF-42 is hereby amended to read as follows:

i 2.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 61, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated in the license. The Corporation shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

The license amendment is effective as of its date of issuance and is to be-implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Suzanne

, Director Project Directorate IV-2 1

Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

March 30, 1993 i

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ATTACHMENT TO LICENSE AMENDMENT NO. 61 FACILITY OPERATING LICENSE NO. NPF-42 DOCKET NO. 50-482 Revised Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.

The revised pages are identified by amendment number and contain marginal lines indicating the area of change.

l The corresponding overleaf pages are also provided to maintain document completeness.

REMOVE INSERT i

I I

II II IV IV V

V XIX XIX l

XXI XXI 1-2 1-2 t

1-3 1-3 1-4 1-4 1

1-5 1-5 1-6 1-6 1-7 1-7 2-2 2-2 2-4 2-4 2-7 2-7 2-8 2-8 2-9 2-9 2-10 2-10 B 2-1 B 2-1

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3/4 1-1 3/4 1-1 3/4 1-2 3/4 1-2 3/4 1-3 3/4 1-3 3/4 1-4 3/4 1-4 3/4 1-5 3/4 1-5 3/4 1-8 3/4 1-8 3/4 1-12 3/4 1-12 3/4 1-14 3/4 1-14 3/4 1-20 3/4 1-20 3/4 1-21 3/4 1-21 3/4 1-22 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-3 3/4 2-3 3/4 2-4 3/4 2-4 3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4 2-7 3/4 2-7 3/4 2-8 3/4 2-8 3/4 2-9 3/4 2-9

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REMOVE INSERT 1

3/4 2-10 3/4 2-10 i

3/4 2-14 3/4 2-14' l

3/4 2-15 3/4 2-15 3/4 2-16 3/4 3-43 3/4 3-43

.i 3/4 7-3 3/4 7-3 l

3/4 9-1 3/4 9-1 3/4 9-16 3/4 9-16 3/4 10-2 3/4 10-2 3/4'10-4 3/4 10-4 i

B 3/4 1-1 B 3/4 1~1 B 3/4 1-2 B 3/4.1-2

-i B 3/4 1-3 B 3/4 1-3 t

B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4.2-2 o

B 3/4 2-3 8 3/4 2-3 B 3/4 2-4 B 3/4 2-4 B 3/4 2-5 l

B 3/4 3-4 B 3/4 3-4 5-6 5-6 5-7 5-7 l

5-8 5-8 6-21 6-21 6-21a.

6-21b i

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DEFINITIONS t

SECTION PAGE

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1.0 DEFINITIONS 1.1 ACTI0N........................................................

1-1 1.2 ACTUATION LOGIC TEST..........................................

1-1 1.3 ANALOG CHANNEL OPERATIONAL TEST...............................

1-1 i

1.4 AXIAL FLUX DIFFERENCE.........................................

1-1 l

1.5 CHANNEL CALIBRATION...........................................

1-1 1.6 CHANNEL CHECK.................................................

1-1 i

1.7 CONTAINMENT INTEGRITY.........................................

1-2 i

1.8 CONTROLLED LEAKAGE............................................

1-2 1.9 CORE ALTERATION...............................................

1-2 j

1.10 CORE OPERATING LIMITS REPORT................................

1-2 1.11 DOSE EQUIVALENT I-131........................................

1-3 1.12 E--AVERAGE DISINTEGRATION ENERGY..............................

1-3 1.13 ENGINEERED SAFETY FEATURES RESPONSE TIME.....................

1-3 i

1.14 FREQUENCY N0TATION...........................................

1-3 1.15 IDENTIFIED LEAKAGE...........................................

1-3 1.16 MASTER RELAY TEST............................................

1-4 1.17 MEMBER (S) 0F THE PUBLIC......................................

1-4 1.18 0FFSITE DOSE CALCULATION MANUAL..............................

1-4 1.19 O PE RAB L E - O PERAB I L ITY.......................................

1-4 1.20 OPERATIONAL MODE - M0DE......................................

1-4 1.21 PHYSICS TESTS................................................

1-4 1.22 PRESSURE BOUNDARY LEAKAGE....................................

1-5

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1.23 PROCESS CONTROL PR0 GRAM......................................

1-5 1.24 PURGE - PURGING..............................................

1-5 1.25 QUADRANT POWER TILT RATI0....................................

1-5 i

1.26 RATED THERMAL P0WER..........................................

1-5 1.27 REACTOR TRIP SYSTEM RESPONSE TIME............................

1-5 1.28 REPORTABLE EVENT.............................................

1-5 1.29 SHUTDOWN MARGIN..............................................

1-6 I

1.30 SITE B0VNDARY................................................

1-6 l

1.31 SLAVE RELAY TEST.............................................

1-6 l

WOLF CREEK - UNIT 1 I

Amendment No. f2,61.

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DEFINITIONS SECTION EAG1 j

i DEFINITIONS (Continued) 1.32 SOURCE CHECK.................................................

1-6 i

t 1.33 STAGGERED TEST BASIS.........................................

1-6 l

t 1.34 THERMAL P0WER................................................

1-6 1.35 TRIP ACTUATING DEVICE OPERATIONAL TEST.......................

1-6 1

1.36 UNIDENTIFIED LEAKAGE.........................................

1-7 1.37 UNRESTRICTED AREA............................................

1-7 l

1.38 VENTILATION EXHAUST TREATMENT SYSTEM.........................

1-7 1.39 VENTING......................................................

1-7 i

1.40 WASTE GAS HOLDUP SYSTEM......................................

1-7 TABLE 1.1 FREQUENCY N0TATION......................................

1-8 TABLE 1.2 OPERATIONAL M0 DES.......................................

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WOLF CREEK - UNIT 1 II Amendment No. 61 e

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

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SECTION PAGE 2.1 SAFETY LIMITS i

2.1.1 REACTOR C0RE................................................

2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE.............................

2-1

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FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION..

2-2

2. 2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0INTS...............

2-3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS....

2-4 l

BASES l

SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................

B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE.............................

B 2-2 0

3 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS...............

B 2-3 9

I WOLF CREEK - UNIT 1 III

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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUTREMENTS SECTION EAQE 3/4.0 APPLICABILITY...............................................

3/4 0-1 i

l 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL

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Shutdown Margin.........................................

3/4 1-1 Moderator Temperature Coefficient........................

3/4 1-3 FIGURE 3.1-1 BOL MODERATOR TEMPERATURE COEFFICIENT VS. POWER LEVEL...........................

3/4 1-5 Minimum Temperature for Criticality......................

3/4 1-6 1

l 3/4.1.2 BORATION SYSTEMS i

Fl ow Path - Sh u td own.....................................

3/4 1-7 l

l Flow Paths - Operating...................................

3/4 1-8 l

Charging Pump - Shutdown.................................

3/4 1-9 Charging Pumps - Operating...............................

3/4 1 1 Borated Water Source - Shutdown..........................

3/4 1-11 I

Borated Water Sources - Operating........................

3/4 1 l 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height.............................................

3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH R0D.........................................

3/4 I-16 Position Indication Systems - Operating..................

3/4 1-17 Position Indication System - Shutdown....................

3/4 1-18 Rod Drop Time............................................

3/4 1-19 Shutdown Rod Insertion Limit.............................

3/4 1-20 t

l Control Rod Insertion Limits.............................

3/4 1-El 1

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i WOLF CREEK - UNIT 1 IV Amendment No. 61 i

i LIMITING CONDITIONS FOR OPERATION AND SURVElltANCE REQUIREMENTS SECTION PAGE 3/4,2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)..............................

3/4 2-1 i

3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F,(X,Y,Z).................

3/4 2-4

+

3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR

-F(X,Y).............................................

-3/4-2-9 n

3/4.2.4 QUADRANT POWER TILT RAT10................................

3/4 2-11 3/4.2.5 DNB PARAMETERS...........................................

3/4 2-14 f

TABLE 3.2-1 DNB PARAMETERS........................................

3/4 2-16 l

3 /4. 3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION......................

3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION...................

3/4 3-2 TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES....

3/4 3-7 r

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................................

3/4 3-9 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION........................................

3/4 3-13 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.....................................

3/4 3-14 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0lNTS......................

-3/4 3-22 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES.............

3/4 3-29 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........

3/4 3-34 3

WOLF CREEK - UNIT 1 V

Amendment No. 61 r

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE INSTRUMENTATION (Continued) 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring for Plant Operations.................

3/4 3-39 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS...............................

3/4 3-40 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS.......................................

3/4 3-42 Movabl e Incore Detectors..................................

3/4 3-43 Seismic Instrumentation...................................

3/4 3-44 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION...................

3/4 3-45 TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION f

SURVEILLANCE REQUIREMENTS..........................

3/4 3-46 Meteorological Instrumentation............................

3/4 3-47 TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION............

3/4 3-48 TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS..........................

3/4 3-49 Remote Shutdown Instrumentation...........................

3/4 3-50 TABLE 3.3-9 REMOTE SNUTDOWN MONITORING INSTRUMENTATION...........

3/4 3-51 TABLE 4.3-6 REMOTE SNUTDOWN MONITORING INSTRUMENTATION j

SURVEILLANCE REQUIREMENTS..........................

3/4 3-52 i

Accident Monitoring Instrumentation.......................

3/4 3-53 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION..................

3/4 3-54 i

TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS..........................

3/4 3-55 Chlorine Detection Systems................................

3/4 3-56 Loose-Part Detection System...............................

3/4 3-57 Radioactive Liquid Effluent Monitoring Instrumentation DELETED TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION DELETED 1

WOLF CREEK - UNIT 1 VI Amendment No. 15, 42

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-DESIGN FEATURES-

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i SECTION PME

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.t 5.1 SITE j

i 5.1.1-EXCLUSION AREA..............................................

5-1" j

i 5.1.2 LOW POPULATION Z0NE..........................................

5-l '

'l 5.1.3 MAPS DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADI0 ACTIVE GASEOUS AND LIQUID EFFLUENTS..............

5-1 f

FIGURE 5.1-1 EXCLUSION AREA.....................................-..

5-2 i

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FIGURE 5.1-2 LOW POPULATION 20NE..................................

5-3 FIGURE 5.1-3 BOUNDARY FOR GASEOUS EFFLUENTS.......................-

5-4 j

FIGURE 5.1-4 BOUNDARY FOR LIQUID EFFUIENTS........................

5-5

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5.2 CONTAINMENT I

5.2.1 CONFIGURATION...............................................

5-1 5.2.2. DESIGN. PRESSURE AND TEMPERATURE.............................

5.

i 5.3 REACTOR CORE I

I 5.3.1 FUEL ASSEMBLIES.............................................

5-6 r

i 5.3.2 CONTROL'R0D ASSEMBLIES......................................

5 l 5.4 REACTOR COOLANT SYSTEM 5.4.1 DESIGN PRESSURE AND TEMPERATURE............................. -

5-6.

j 5.4.2 V0LUME.................................................

5-6 5.5 METEOROLOGIC AL TOWER LOC ATION.................................

5-6 j

i 5.6 FUEL STORAGE-t i

f 5.6.1 CRITICALITY..............................................

5-7 5.6.2 DRAINAGE....................................................

5-7' l

5.6.3 CAPACITY....................................................

5-7L

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l 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT...........................

5-7 j

TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMIT...................

5-9 l

k WOLF CREEK UNIT I XIX Amendment: No. 61.

1 r

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ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 _ RESPONSIBILITY................................................

6-1 6.2 ORGANIZATION 5.2.1 ONSITE AND OPERATING CORPORATION ORGANIZATION...............

6-1 6.2.2 UNIT STAFF..................................................

6-1 i

FIGURE 6.2-1 DELETED..............................................

6-3 FIGURE 6.2-2 DELETED..............................................

6-4 j

TABLE 6.2-1 MINIMUM SHIFT CREW COMP 05ITION........................-

6-5 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

Function....................................................

6-6 l

Composition..................................................

6-6 Responsibilities............................................

6-6 Records.....................................................

6-6 6.2.4 SHIFT TECHNICAL ADVI50R.....................................

6-6 6.3 UNIT STAFF 0UALIFICATIONS.....................................

6-6 r

6.4 TRAINING......................................................

6-7 6.5 REVIEW AND AUDIT 6.5.1 PLANT SAFETY REVIEW COMMITTEE (PSRC)

Function....................................................

6-7 Compo s i ti o n.................................................

6-7 Alternates..................................................

6-7

^

Meeting Frequency...........................................

6-8 i

Quorum......................................................

6-8 Responsibilities............................................

6-8 Records.....................................................

6-9 h

WOLF CREEK - UNIT 1 XX Amendment No.24

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ADMINISTRATIVE CONTROLS SECTION PAGE 6.5.2 NUCLEAR SAFETY REVIEW COMMITTEE (NSRC)

Function...................................................

i *

' Composition.......................................

J-10 Alternates..................................................

6-10 Consultants.................................................

6-10 Heeting Frequency............

6-10 Quorum......................................................

6-10 Review......................................................

6-11 Audits......................................................

6-11 Records.....................................................

6-12 l

6.6 REPORTABLE EVENT ACTI0N.......................................

6-13 6.7 SAFETY LIMIT VIOLATION........................................

6-13 0

6.8 PROCEDURES AND PR0 GRAMS.......................................

6-13 6.9 REPORTING REOUIREMENTS i

6.9.1 ROUTINE REP 0RTS.............................................

6-17 Startup Report..............................................

6-17 An n u a l Re p o r t s..............................................

6-17 Annual Radiological Environmental Operating Report..........

6-18 Semiannual Radioactive Effluent Release Report..............

6-19 Monthly Operating Report....................................

6-20 Core Operating Limits Report...............................

6-21 6.9.2 SPECIAL REP 0RTS.............................................

6-21a l

6.10 RECORD RETENTION.............................................

6-21a t

h WOLF CREEK - UNIT 1 XXI Amendment No. 52, 61 i

i ADMINISTRATIVE CONTROLS SECTION PAGE 6.11 RADIATION PROTECTION PR0 GRAM......................................

6-23 6.12 HIGH RADIATION AREA...............................................-

6-23 6.13 PROCESS CONTROL PROGRAM (PCP)...

6-24 6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM)............................

6-24 i

6.15 DELETED t

WOLF CREEK - UNIT 1 XXII Amendment No. 42

1.0 DEFINITIONS i

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The defined terms of this section appear in capitalized type and are applicable throu;;hout these Technical Specifications.

ACTION 1.1 ACTION shall be that part of a Technical Specification which prescribes remedial measures required under designated conditions.

ACTUATION LOGIC TEST 1.2 An ACTUATION LOGIC TEST shall be the application of various simulated input combinations in conjunction with each possible interlock logic state and verification of the required logic output.

The ACTUATION LOGIC TEST shall include a centinuity check, as a minimum, of output devices.

ANALOG CHANNEL OPERATIONAL TEST 1.3 An ANALOG CHANNEL OPERATIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock and/or trip functions. The ANALOG CHANNEL OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, inter-lock and/or Trip Setpoints such that the Setpoints are within the required.

range and accuracy.

AXIAL FLUX DIFFERENCE 1.4 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector.

CHANNEL CALIBRATION 1.5 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such t.et it responds within the seguired range and accuracy to known values of inpat. The CHANNEL CALIBRATION shall encompass the entire. channel including the sensors and alarm, interlock and/or trip functions and may be performed by any. series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK

1. 6 A CHANNEL CHECK shall be the qualitative assessment of channel oehavior during operation by observation.

This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from incependent instrument channels 5

measuring the same parameter.

I WOLF CREEK - UNIT 1 1-1 F

DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a.

All penetrations required to be closed during accident conditions are either-1)

Capable of being closed by an OPERABLE containment automatic isolation valve system, or 2)

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.

b.

All equipment hatches are closed and sealed, c.

Each air lock is in compliance with the requirements of Specification 3.6.1.3, d.

The containment leakage rates are within the limits of l

Specification 3.6.1.2, and e.

The sealing mechanism asscciated with each penetration (e.g.,

I welds, bellows, or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE l

1.8 CONTROLLED LEAKAGE shall be that seal water flow from the reactor coolant pump seals.

CORF ALTERATION 1.9 CORE ALTERATION shall be the movament or manipulation of any component within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT L

1.10 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document l

that provides core operating limits for the current operating reload cycle.

These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9.

Plant operation within these operating limits is addressed in individual Specifications.

i i

n WOLF CREEK - UluT 1 1-2 Amendment No. 61 i

a DEFINITIONS DOSE E0VIVALENT I-131 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie /

l gram) which alone would produce the same thyroid dose as the quantity and iso-topic mixture of I-131, I-132,- I-133, I-134, and I-135 actually. present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

E - AVERAGE DISINTEGRATION ENERGY 1.12 E shall be the average (weighted in proportion to the concentration of l

each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half-lives greater than 15 minutes, making up at least 95% of the total noniodine activity in the coolant.

ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time

.l interval from when the monitored parameter exceeds its ESF Actuation Setpoint 4

at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FRE0VENCY NOTATION l.14 The FREQUENCY NOTATION specified for the performance of Surveillance l

Requirements shall correspond to the intervals defined in Table 1.1.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEA +1? 0:all be:

l l

a.

Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or b.

Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or

{

c.

Reactor Coolant System leakage through a steam generator to the Secondary Coolant System.

)

i WOLF CREEK - UNIT 1 1-3 Amendment No. 61 i

DEFINITIONS MASTER RELAY TEST 1.16 A MASTER RELAY TEST shall be the energization of each master relay and f

verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

MEMBER (S) 0F THE PUBLIC 1.17 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the pl ant.

OFFSITE DOSE CALCULATION MANUAL 1.18 The OFFSITE DOSE CALCULATION MANUAL (0DCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semiannual Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.7.

OPERABLE - OPERABILITY 1.19 A system, subsystem, train, component or device shall be OPERABLE or l

have OPERABILITY when it is capable of performing its specified function (s),

and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s).

OPERATIONAL MODE - MODE 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.

PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental l

nuclear characteristics of the core and related instrumentation: (1) described in Chapter 14.0 of the FSAR, or (2) authorized under'the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

WOLF CREEK - UNIT 1 1-4 Amendment No. 61 t

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DEFINITIONS PRESSURE BOUNDARY LEAKAGE r

1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube lI leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.

i PROCESS CONTROL PROGRAM i

1.23 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, l'

sampling, analyses, tests, and determinations to be made to ensure that the I

processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

PURGE - PURGING l;

1.24 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

OUADRANT POWER TILT RATIO 1.25 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore l:!

detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining

+

three detectors shall be used for computing the average.

RATED THERMAL POWER 5

1.26 RATED THERMAL POWER shall be a total core heat transfer rate to the l

reactor coolant of 3411 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME l.27 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from l'

when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE EVENT j

1.28 A REPORTABLE EVENT shall be any of those conditions specified in l

Section 50.73 to 10 CFR Part 50 I

WOLF CREEK - UNIT 1 1-5 Amendment No. /2, 61 i

DEFINITIONS SHUTDOWN MARGIN 1.29 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which l

the reactor is subcritical or would be subcritical from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod c1cster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY 1.30 The SITE B0UNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

5 SLAVE RELAY TEST 1.31 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.

SOURCE CHECK 1.32 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

STAGGERED TEST BASIS 1.33 A STAGGERED TEST BASIS shall consist of:

a.

A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals, and b.

The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

THERMAL POWER 1.34 THERMAL POWER shall be the total core heat transfer rate to the reactor coolant.

TRIP ACTUATING DEVICE OPERATIONAL TEST 1.35 A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the Trip Actuating Device and verifying OPERABILITY of elarm, interlock and/or trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary, of the Trip Actuating Device such that it actuates at the required Satpoint within the required accuracy.

WOLF CREEK - UNIT 1 1-6 Amendment No. #2, 61

DEFINITIONS UNIDENTIFIED LEAKAGE I.36 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

UNRESTRICTED AREA I.37 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE B0UNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

VENTILATION EXHAUST TREATMENT SYSTEM 1.38 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radiciodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through -

charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents.

Engineered Safety Features (ESF) Atmospheric Cleanup Systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING 1.39 VENTING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner _that rep'lacement air or gas is not provided or required during VENTING.

Vent, used in system names, does not imply a VENTING process.

WASTE GAS HOLDUP SYSTEM 1.40 A WASTE GAS HOLDUP SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting Reactor Coolant System off-gases from the Reactor Coolant System and providing for delay or holdup for the purpost of reducing the total radioactivity prior to release to the environment.

-i

}

I P

h WOLF CREEK - UNIT 1 1-7 Amendment No. 61

\\

l

l l

TABLE 1.1 j

FREQUENCY NOTATION i

NOTATION FREQUENCY l

S At least'once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

W At least once per 7 days.

l M

At least once per 31 days.

Q At least once per 92 days.

SA At least once per 184. days.

R At least once per 18 months.

S/U Prior to each reactor startup.

l N.A.

Not applicable.

P Completed prior to each release.

P t

WOLF CREEK - UNIT 1 1-8

l 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, operating loop coolant temperature (T,yg) pressurizer pressure, and t shall not exceed the limits shown in Figure 2.1-1 for four loop operation.

APPLICABILITY: MODES 1 and 2.

f ACTION:

Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pres-surizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY:

MODES 1, 2, 3, 4, and 5.

ACTION:

MODES I and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within I hour, and comply with the requirements of Specification 6.7.1.

MODES 3, 4, and 5:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, I

reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1.

l l

WOLF CREEK - UNIT 1 2-1

i

.-f e

.f 680 l

-1 l

I UNACCEPTABLE i

660 N

OPERATION l

N I

2400 PSIA -

N N

,y N

N i

N

\\

g LIJ 2000 PSIA k

f N

2250 PSIA -+

-\\

m b 620

\\

g N

N N\\

N,

\\\\

.l p

190d PSIA

_N N

\\\\

l y3 O

\\

l

600

^ x\\ \\\\

\\\\ \\,

i ACCEPTABLE

.g g OPERATION

\\\\

580

\\N

.\\

i i

560 O.0 0.1 0.2 0.3 0.4 0.5 0.6

. 0.7 0.8 0.9 1.0 1.1 1.2 3

FRACTION OF RATED THERMAL-POWER i

FIGURE 2.1-1 j

REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION l

a WOLF CREEK. UNIT 1 2-2 Amendment No. 61 i

g TABLE 2.2-1 E

REACTOR TRIP SYSTEM INSTRUNENTATION TRIP SETPOINTS 9

SENSOR n

TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA)

Z (S)

TRIP SETPOINT ALLOWABLE VALUE 1.

Manual Reactor Trip N.A.

N.A.

H.A.

H 2.

Power Range, Neutron Flux a.

High Setpoint 7.5 4.56 0

1109% of RTP*

1112.3% of RTP*

b.

Low Setpoint 8.3 4.56 0

125% of RTP*

128.3% of RTP*

3.

Power Range, Neutron Flux, 2.4 0.5 0

<4% of RTP* with

<6.3% of RTP* with High Positive Rate i time constant i time constant 12 seconds 12 seconds 4.

Power Range, Neutron Flux, 2.4 0.5 0

<4% of RTP* with

<6.3% of RTP* with High Negative Rate i time constant i time constant m

A 12 seconds 12 seconds 5.

Intermediate Range, 17.0 8.41 0

125% of RTP*

135.3% of RTP*

Neutron Flux 8

5 6.

Source Range, Neutron Flux 17.0 10.01 0

110 cps 11.6 x 10 cps 7.

Overtemperature AT 7.3 5.11 2.42 See Note 1 See Note 2 l

[

8.

Overpower AT 4.60 2.93 0.14 See Note 3 See Note 4 9.

Pressurizer Pressure-Low 3.7 0.71 2.49 11915 psig 11906 psig 10.

Pressurizer Pressure-High 7.5 0.71 2.49 12385psig 12400psig

(

11.

Pressurizer Water Level-High 8.0 2.18 1.96

$92% of instrument 193.9% of instrument span span n*

RTP = RATED THERMAL POWER

- Loop design flow = 93,600 gpm b

e

g TABLE 2.2-1 (Cgntinued)

G n

TABLE NOTATIONS NOTE 1:

OVERTEMPERATURE AT h

ATff)2 (1

TsS} I O o (Kt-K2 ff

{

[T (1 f

3) - T'] + Ks(P P') - f (^I)I 1

Q Where:

AT Measured AT;

=

f jjf

= Lead-lag compensator on measured AT; Time constants utilized in lead-lag compensator for AT, it = 6 s, T, 13

=

1 T2 = 3 s; 1

TsS Lag spensator on measured AT;

=

Time constant utilized in the lag compensator for AT, 13 = 2 s; T3

=

AT, Indicated AT at RATED THERMAL POWER;

=

Kg 1.10;

=

K2 0.0137/'F;

=

f jgf

= The function generated by the lead-lag compensator for T,yg e

p dynamic compensation; g

k Time constants utilized in the lead-lag compensator for T,y9, 14 = 16 s, T, Ts

=

4 a

ts = 4 s; x

P T

=

Average temperature, 'F; I

y 7, ts5 Lag compaator on measund T,yg;

=

E Time constant utilized in the measured T,yg lag compensator, is = 0 s; Ts

=

4

4 5

t TABLE 2.2-1 (Continued) 4 9

j.

TABLE NOTATIONS-(Continued) i %

NOTE 1:. (Continued)-

n L

T' 588.5'F (Nominal T,yg.at RATED THERMAL POWER);

siip Ks 0.000671;

=

P

=

Pressurizer pressure, psig; P'

2235 psig (Nominal RCS operating pressure);

4

=

S-

- =. Laplace transform operator, s 1; and fg(AI).is a function of the indicated difference.between top and bottom detectors of the power-range neutron ion-chambers;.with gains to be-selected based on measured instrument response during plant STARTUP tests such that:

-ty

'(1) for qt ~L9b between -25% and + N, f (AI) = 0, where qt.and qb are percent' t

RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt +.9 is-b

. total THERMAL POWER in percent of RATED-THERMAL. POWER; (11)' for each percent that'the magnitude of qt exceeds;-25% the AT Trip Setpoint l

~9b shall be' automatically reduced by 1.8%: of its value at RATED THERMAL POWER;.and l

(111) for each percent that.theLaagnitude of q qb exceeds +N, the AT Trip Setpoint

~

g lf shall be automatically reduced by 1.384% of its value at RATED THERMAL ~ POWER.

I

~

E.

= NOTE 2:

The channel's-maximum Trip.Setpoint shall not' exceed its computed Trip.Setpoint by more than g

1.6%.of AT span.

,N -

-[

i L

. g:

,m_

ii- -- - -,- m

_.,_...-_r-m---..'..~.,..'..~,.m

---.-,----4.+...

-..,.,-...m~-.,-.e+.-,,..+,m

,=~..m.v

.c+-,-

TABLE 2.2-1 (Continued)

Q TABLE NOTATIONS (Continued) n NOTE 3:

OVERPOWER AT ATh fy[73

$ AT, (K4-Ks(1Yz,3 If5 T - Ks [T ff5 - T"] - f (al))

)

3 l

18 l

18 2

E ti Where:

AT

=

. Measured AT; 1

I Lead-lag compensator on measured AT;

=

Time constants utilized in lead-lag compensator for AT, ta = 6 s,12 = 3 s; Ts. In

=

1f 5

Lag compensator on measured AT;

=

13 Time constant utilized in the lag compensator for AT, ta = 2 s; m

ta

=

AT, Indicated AT at RATED THERMAL POWER;

=

K4 1.10;

=

0.02/"F for increasing average temperature and 0 for decreasing average Ks

=

. temperature; k

3j'fys The function generated by the rate-lag compensator for T dynamic

=

compensation; ava Time constant utilized in the rate-lag compensator for T,yg, ty = 10 s;

=

ty 1

1 + 1s5 Lag compensator nn measured T,yg;

=

E Ts Time constant utilized in the measured T,yg lag compensator, is = 0 s;

=

t g-

m

_ TABLE 2.2-1 (Continued) i 5

r-L m-TABLE NOTATIONS-(Continued) g.

p NOTE 3:

(Continued) n.

Ks 0.00128/*F for T > T" and Ks = 0 for T $ T";

=

.g

t T

Average temperature, 'F;

=

L w

T" Indicated T at RATED. THERMAL POWER (Calibration temperature for AT

=

av9 instrumentation, 5 588.5'F);

5

=

Laplace transfom operator, s 1; and f (AI) ~

0 for all AI.

=

I NOTE 4:

The channel's' maximum Trip Setpoint shall not exceed its computed Trip Setpotut by more than-T 1.67%,-AT: span.

M t

't l'

x y

' D W.

L m.

. w

+..'+.,va--4y.

.w..v.<y

,,,,, +,.~*,w,*w--

- ryv-.,-%.w.-wi,y+ ~w - e U-w m,... - c, 3..,*,--,w--+-.,w,'.--+w'.,,wE-*Er.

-v~.-o-v.-*

mw,~++.4...w.v-r,.-ww-.,

e.-.,vre n-

---&--,w.+-....

rv

,4-

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE P

The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is pre-vented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the DNBR correlations. DNBR correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, and is indicative of the margin to DNB.

The DNB design basis is as follows: there must be at least a 95 percent probability that the minimum DNBR of the limiting rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation as specified in the CORE OPERATING LIMITS REPORT (COLR). The correlation.DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit.

For plant conditions which fall outside the range of applicability of the DNB correlation above, the W-3 correlation is used.

In addition, DNB margin is maintained by performing safety analyses to a higher value than the correlation limit, called the safety analysis limit DNBR. The margin between the safety analysis limit DNBR and the correlation limit DNBR is used to cover known DNBR penalties and provide margin for design flexibility. The safety analysis limit DNBR is specified in the COLR.

l The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the applicable safety analysis limit DNBR, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

These curves are based on the design F specified in the COLR and a u

reference cosine with a peak of 1.55 for axial power shape.

i WOLF CREEK - UNIT 1 B 2-1 Amendment No. 23. EI, 61

~

SAFETY LIMITS BASES 2.1.1 REACTOR CORE (Continued) l These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum alloweble control rod insertion assuming the axial power imbalance is within the limits of the f (AI) function of the Overtemperature trip.

k' hen the axial power imbalance 2is not within the tolerance, the axial power imbalance effect on the Over-temperature AT trips will reduce the Setpoints to provide protection consistent with core Safety Limits.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System (RCS) from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor vessel, pressurizer, and the RCS piping and valves are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated Code requirements.

The entire RCS is hydrotested at greater than or equal to 125% (3110 psig) of design pressure, to demonstrate integrity prior to initial operation.

k'OLF CREEK - UNIT 1 B 2-2 Amendment No. SI

i 3 /4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORAT10N CONTROL SHUTDOWN MARGIN j

LIMITING CONDITION FOR OPERATION

[

t 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.3% Ak/k for -

f four loop operation.

APPLICABILITY: MODES 1, 2*, 3, 4, and 5.

/LcJJ03:

With the SHUTDOWN MARGIN less than 1.3% Ak/k, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing

.j greater than or equal to 7000 ppm boron or equivalent-until the required SHUTDOWN MARGIN is restored.

)

SURVEILLANCE RE0VIREMENTS l

4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal i

to 1.3% Ak/k-a.

Within I hour after detection of an inoperable control rod (s) and-i at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while-the rod (s) is inoperable.

If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control. rod (s);

j b.

When in MODE 1 or MODE 2 with K greater. than or equal-to 1 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verilSing that control bank withdrawal is within the limits of Specification 3.1.3.6;

i

.c.

When in MODE 2 with K less than 1, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months _ prior to achieving reactor. criNcality by verifying that the predicted critical control rod position is within the limits of Specification.3.1.3.6; d.

Prior to initial operation above-5% RATED THERMAL POWER after each i

fuel loading,-by consideration of.the factors of Specification-4.1.1.1.le. below, with the control banks at. the maximum insertion limit of Specification 3.1.3.6; and i

1 i

See Special Test Exception Specification 3'10.1.

WOLF CREEK - UNIT 1 3/4 1-1 Amendment No. 61

)

t

REACTIVITY CONTROL SYSTEMS L

SURVEILLANCE REOUIREMENTS (Continued) e.

When in MODE 3, 4, or 5, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors-1)

Reactor Coolant System boron concentration, j

2)

Control rod position, 3)

Reactor Coolant System average temperature,

{

4)

Fuel burnup based on gross thermal energy generation, 5)

Xenon concentration, and' 6)

Samarium concentration.

f 4.1.1.1.2 The overall core reactivity balance shall be compared to predicted i

values to demonstrate agreement within i 1% ok/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification 4.1.1.1.le. above..The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD after each fuel loading.

r l

5 WOLF CREEK - UNIT 1 3/4 1-2 Amendment No. 61 j

r L

REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (HTC) shall be:

a.

Less positive than the limits specified in Figure 3.1-1 for the all rods withdrawn, beginning of cycle life (BOL), THERMAL POWER condition, b.

Less negative than the E0L limit specified in the COLR for the all rods withdrawn, end of cycle life (EOL), RATED THERMAL POWER condition.

APPLICABILITY: Specification 3.1.1.3a. - MODES I and 2#*.

Specification 3.1.1.3b. - MODES 1, 2, and 3#.

ACTION:

a.

With the MTC more positive than the limit of Specification 3.1.1.3a, above, operation in MODES 1 and 2 may proceed provided:

1.

Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than the limits specified in Figure 3.1-1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6; 2.

The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the HTC has been restored to within its limit for the all rods withdrawn condition; and 3.

A Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

b.

With the MTC more negative than the EOL limit specified in the COLR, be in HOT SHilTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With K, greater than or equal to 1.

g

See Special Test Exception Specification 3.10.3.

WOLF CREEK - UNIT 1 3/4 1-3 Amendment No. 61

^

t i

i l

REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE0VIREMENT4 4.1.1 -. 3 The MTC shall be determined to be within its limits during each fuel cycle as follows:

i a.

The MTC shall be measured and compared to the BOL hot zero power limit specified in Figure 3.1-1 prior to initial operation above 5%

i of RATED THERMAL POWER, after each fuel loading; and b.

The MTC shall be measured at any THERMAL POWER and compared to the 300 ppm surveillance limit specified in the COLR (all rods l

t withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm.

In the event this comparison indicates the MTC is more negative than the-i 300 ppm surveillance limit specified in the COLR, the MTC shall be-remeasured, and compared to the EOL MTC limit, at least once per 14 i

EFPD during the remainder of the fuel cycle.

s t

f i

i 4

s I

4 f

WOLF CREEK - UNIT 1 3/4 1-4 Amendment No. 61

i MTC (pcm/deg. F)

ARO at BOL-8 LHACCEPTABLE OPERATION 6.0, 70:

~

~~~~~~--

6 i

ACCEPTABLE OPERATION 4

- ~ ~ ~ ~ ~ ~ - -

2 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

~!

h I

I i

f I

I I

I l-O O

-10 20-30 40 50 60 70- 90 100' i

i

% of Rated Thermal Power l

i

?

Figure 3.1-1 BOL MODERATOR TEMPERATURE COEFFICIENT VS. POWER LEVEL.

1 l

.l WOLF CREEK - UNIT 1 3/4 1-5 Amendment No. 61

REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY j

i LIMITING CONDITION FOR OPERATION 3.1.1.4 shall be greater than or equal to 551*F.The Reactor Coolant System lowest opera APPLICABILITY: MODES 1 and 2#*.

ACTION:

i With a Reactor Coolant System operating loop temperature (T,yg) less than

[

551 F, restore T,yg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes.

SURVEILLANCE REQUIREMENTS 4.1.1. 4 The Reactor Coolant System temperature (T"V9) shall be determined to be greater than or equal to 551*F:

Within 15 minutes prior to achieving reactor criticality, and a.

b.

At least once per 30 minutes when the reactor is critical and the Reactor Coolant System T,yg is less than 561*F with the T,yg-T ref Deviation Alarm not reset.

With K,ff greater than or equal to 1.

"See Special Test Exception Specification 3.10.3.

WOLF CREEK - UNIT 1 3/4 1-6

REACTIVITY CONTROL SYSTEMS 3/4.1.2 BORATION SYSTEMS FLOW PATH - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.1 As a minimum, one of the following boron injection flow paths shall be OPERABLE.and capable of being powered from an OPERABLE emergency power source:

a.

A flow path from the Boric Acid Storage System via a boric acid transfer pump and a centrifugal charging pump to the Reactor Coolant System if the Boric Acid Storage System is OPERABLE as given in Specification 3.1.2.5a. for MODES 5 and 6 or as given in Specification 3.1.2.6a. for MODE 4; or b.

The flow path from the refueling water storage tank via a centrifugal charging pump to the Reactor Coolant System if the refueling water storage tank is OPERABLE as given in Specification 3.1.2.5b. for MODES 5 and 6 or as given in Specification 3.1.2.6b. for MODE 4.

APPLICABILITY:

MODES 4, 5, and 6.

ACTION:

With none of the above flow paths OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS i

4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABLE at least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

a

?

WOLF CREEK - UNIT 1 3/4 1-7

REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE:

a.

The flow path from the Boric Acid Storage System via a boric acid transfer pump and a centrifugal charging pump to the Reactor Coolant System, and b.

Two flow paths from the refueling water storage tank via centrifugal charging pumps to the Reactor Coolant System.

APPLICABILITY: MODES 1, 2, and 3.*

ACTION:

With only one of the above required baron injection flow paths to' the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1.3% Ak/k at 200'F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two flow paths to j

OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:

a.

At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position; b.

At least once per 18 months during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection test signal; and c.

At least once per 18 months by verifying that the flow path required by Specification 3.1.2.2a. delivers at least 30 gpm to the Reactor Coolant System.

The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODE 3 for the centrifugal charging pump declared inoperable pursuant to Specification 4.1.2.3.2 provided that the centrifugal charging pump is restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or prior to the temperature of one or more of the RC3 cold legs exceeding 375'F, whichever comes first.

WOLF CREEK - UNIT 1 3/4 1-8 Amendment No. 61

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCE - SHUTOOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE:

a.

A Boric Acid Storage System with:

1)

A minimum contained borated water volume of 2958 gallons, 2)

Between 7000 and 7700 ppe of boron, and 3)

A minimum solutien temperature of 65'F.

b.

The refueling water storage tank (RWST) with:

1)

A minimum contained borated water volume of 55,416 gallons, 2)

A minimum boron concentration of 2400 ppm, and

[

3)

A minimum solution temperature of 37'F.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS 4.1.2.5 The above -required borated water source shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1)

Verifying the boron concentration of the water, 2)

Verifying the contained borated water volume, and 3)

Verifying the Boric Acid Storage Systen solution temperature when it is the source of borated water.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when it is the source of borated water and the outside air temperature is less than 37'F.

WOLF CREEK - UNIT 1 3/4 1 Amendment No. 23

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I REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 1

3.1.2.6 As a minimum, the following borated water sources shall be OPERABLE as required by Specification 3.1.2.2 for MODES 1, 2, and 3 and one of the -

following borated water sources shall be OPERABLE as required'by Specifica-l tion 3.1.2.1 for MODE 4:

a.

A Boric Acid Storage System with:

[

I 1)

A minimum contained borated water volume of 17,658 gallons, i

2)

Between 7000 and 7700 ppm of boron, and f

3)

A minimum solution temperature of 65'F.

j b.

The refueling water storage tank (RWST) with:

j 1)

A minimum contained borated water volume of 394,000 gallons,

-l 2)

Between 2400 and 2500 ppm of boron, 3)

A minimum solution temperature of 37*F, and 4)

A maximum solution temperature of 100*F.

j APPLICABILITY: H0 DES 1, 2, 3, and 4.

ACTION:

B a.

With the Boric Acid Storage System inoperable and being used as one i

of the above required borated water sources in MODE 1, 2 or 3, restore the storage system to OPERABLE' status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be i

in at least HOT STANDBY within the next.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a-

-l SHUTDOWN MARGIN equivalent to at least 1.3% Ak/k at 200*F; restore l

the Boric Acid Storage System to OPERABLE status within the:next l

7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .-

i b.

With the RWST inoperable in MODE 1, 2, or 3, restore the tank.to j

OPERABLE status within I hour or be in at least HOT STANDBY within 1

the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and. in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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c.

With no borated water source. OPERABLE in. MODE 4, restore one borated water source to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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i WOLF CREEK

-UNIT 1 3/4 1-12 Amendment No. 23, 61 i

REACTIVfTY CONTROL SYSTEMS 4

SURVEILLANCE REOUIREMENTS 4.1.2.6 Each required borated water source shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1)

Veri.'ying the boron concentration in the water, 2)

Verifying the contained borated water volume of the water source, and 3)

Verifying the Boric Acid Storage System solution temperature when it is the source of borated water.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when the outside air temperature is either less than 37'F or greater than 100*F.

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t WOLF CREEK - UNIT 1 3/4 1-13

.9.-

REACTIVITY CONTROL SYSTEMS

y 3/4.I'3 MOVABLE CONTROL ASSEMBLIES i

GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All full-length ' shutdown and control rods shall be OPERABLE and positioned within i 12 steps (indicated position) of.their group step counter i

demand position, j

APPLICABILITY: MOCES~1* and 2*.

l ACTION: The ACTION to be taken is based on the cause of inoperability of control rods as follows:

)

i ACTION:

More Than:

CAUSE OF IN0PERABILITY One Rod One Rod a)

Immovable as a result of excessive (1)

(1)-

friction or mechanical interference or known to be untrippable.

I b) Misaligned from its group step

'(3)'

(2)

{

counter demand height ~or from any other rod in its group by more than i 12 steps (indicated position).

j c)

Inoperable due to a rod control urgent

-(4)

.1 failure alarm or other electrical i

problem in the rod control system, but trippable.

~

t ACTION 1 - Determine that the SHUTDOWN MARGIN. requirement of Specification 1

3.1.1.1 is satisfied within I hour and be in HOT STANDBYTwithin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 2 - Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 3 - POWER OPERATION may~ continue provided that within I hour:

1.

The rod is restored to OPERABLE status within the above' l

alignment requirements, or 2.

The rod is declared inoperable and the remainder of. the rods _

l in the group with the inoperable rod are aligned. to within;i 12 steps of the inoperable rod while paintaining the rod.

sequence and insertion limits of Specification 3.1.3.6.

The.

THERMAL POWER level shall be; restricted pursuant' to Specification 3.1.3.6'during subsequent: operation, or-t

5ee Special Test Exceptions Specific.ations 3.10.2 and 3.10.3.

l WOLF-CREEK - UNIT 1 3/4 1-14 Amendment No. 27, M, 61 l

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i REACTIVITY CONTROL SYSTEMS j

ROD DROP TIME i

LIMITING CONDITION FOR OPERATION j

3.1.3.4 The individual full-length shutdown and control rod drop time from the physical fully withdrawn position shall be less than or equal to 2.7 seconds l

from beginning of decay of stationary gripper coil voltage to dashpot entry with:

T,yg greater than or equal to 551'F, and a.

b.

All reactor coolant pumps operating.

APPLICABILITY: MODES 1 and 2.

ACTION:

a.

With the rod drop time of any full-length rod determined to exceed the above limit, restore the rod drop time to within the above limit i

prior to proceeding to MODE 1 or 2.

b.

With the rod drop times within limits but determined with three reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to 66% of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full-length rods shall be demonstrated through measurement prior to reactor criticality:

a.

For all rods following each removal of the reactor vessel head, b.

For specifically affected individual rods following any maintenance on or modification to the Control Rod Drive System which could affect the drop time of those specific rods, and c.

At least once per 18 months.

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WOLF CREEK - UNIT 1 3/4 1-19 Amendment No. 32, 47

i REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT LIMITING CONDITION FOR OPERATION I

3.1.3.5 All shutdown rods shall be limited in physical insertion as specified-in the CORE OPERATING LIMITS REPORT (COLR).

j APPLICABILITY: MODES 1* and 2*#.

ACTION:

i With a maximum of one shutdown rod inserted beyond the insertion limit specified in the COLR, except for surveillance testing pursuant to Specification 4.1.3.1.2, within I hour either:

a.

Restore the rod to within the insertion limit specified in the COLR, or b.

Declare the rod to be inoperable and apply Specification 3.1.3.1.

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SURVEILLANCE REOUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be within the insertion limit specified in the COLR:

t Within 15 minutes prior to withdrawal of any <ods in Control Bank A, a.

B, C, or D during an approach to reactor criticality, and b.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

s

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

With K,,, greater than or equal to 1.

[

i WOLF CREEK - UNIT 1 3/4 1-20 Amendment No. 61

t 4

REACTIVITY CONTROL SYSTEMS CONTROL R0D INSERTION LIMITS LIMITING CONDITION FOR OPERATION 1

3.1.3.6 The control banks shall be limited in physical insertion.as specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY: MODES 1* and 2*#.

ACTION:

l With the control banks inserted beyond the insertion limits specified in the COLR, except for surveillance testing pursuant to Specification 4.1.3.1.2:

a.

Restore the control banks to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or b.

Reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to less than or equal to that fraction of RATED THERMAL POWER which is allowed _by the bank l

position using the insertion limits specified in the COLR, or l

c.

Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REOUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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1

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

fWith K,,, greater than or equal to 1.

WOLF CREEK - UNIT 1 3/4 1-21 Amendment No. 61

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3/4.2' POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD) l 1

LIM [IJ3GCONDITIONFOROPERATION l

3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the allowed operational space specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY: MODE I above 50 PERCENT RATED THERMAL POWER *.

[

ACTION:

[

a.

With the indicated AFD outside of the limits specified in the COLR, 1.

Either restore the indicated AFD to within the limits specified in the COLR within 15 minutes, or l

2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER f

within 30 minutes and reduce the Power Range Neutron Flux -

High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

THERMAL POWER shall not be is?reased above 50% of RATED THERMAL POWER unless the indicated afb is within the limits specified in the COLR.

F

See Special Test Exception 3.10.2.

WOLF CREEK - UNIT 1 3/4 2-1 Amendment No. Y, 61 l

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION e

1 SURVEILLANCE REOUIREMENTS l.

I i

4. 2.1.1 The indicated AFD shall be determined to be within its limits during i

POWER OPERATION above 50% of RATED THERKAL POWER by:

l a.

Monitoring the indicated AFD for each OPERABLE excore channel:

1)

At least once per 7 days when the AFD Monitor Alarm is OPERABLE, l

and j

2)

At least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring i

the AFD Monitor Alarm to OPERABLE status.

b.

Monitoring and logging the indicated AFD for each OPERABLE excore.

l channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is l

inoperable.

The logged values of the indicated AFD shall be assumed i

to exist during the interval preceding each logging.

l 4.2.1.2 The indicated AFD shall be considered outside of its limits when at least two OPERABLE excore channels are indicating the AFD to be outside the.

limits.

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i WOLF CREEK - UNIT 1 3/4 2-2 Amendment No. 1

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1 WOLF CREEK - UNIT 1 3/4 2-3 Amendment No. 7,61 i

i POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F.JX.Y.Z)

LIMITING CONDITION FOR OPERATION 3.2.2 F,(X,Y,Z) shall be limited by the following relationships:

i F,"(X,Y,Z) s [F,"] [K(Z)] for P > 0.5, and P

F,"(X,Y,Z) s [F,"] [K(Z)] for P s 0.5.

0.5 Where

F,"(X,Y,Z) - the measured heat flux hot channel factor, F,"(X,Y,Z), increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty, F,"'"

- the F Limit at RATED THERMAL POWER (RTP),

i as specified in the CORE OPERATING LIMITS REPORT (COLR),

P

= THERMAL POWER

, and RATED THERMAL POWER K(Z)

= the normalized F,(X,Y,Z) limit as a function of core height, as specified in the COLR.

APPLICABILITY: MODE 1.

ACTION:

With F,(X,Y,Z) exceeding its limit:

Reduce THERMAL POWER at least 1% for each 1% F "(X,Y,Z) exceeds the l

a.

t limit within 15 minutes and similarly reduce tie Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and b.

Control the AFD to within new AFD limits which are determined by reducing the allowable THERMAL POWER at each point along the AFD 1

limit lines of Specification 3.2.1 at least 1% for each 1%

F,"(X,Y,Z) exceeds the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and declare the AFD monitor alarm inoperable until the AFD alarm setpoints are reset to the modified limits; and c.

POWFR OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWEP, OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% F,"(X,Y,Z) exceeds the limit; and WOLF CREEK - UNIT 1 3/4 2-4 Amendment No. 61

h POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (X.Y.7)

LIMITING CONDITION FOR OPERATION (Continued) d.

Identify and correct the cause of the cut-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION a., above; THERMAL POWER may then be increased provided F,(X,Y,Z) is demonstrated through incore mapping to be within its s

limit.

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t WOLF CREEK - UNIT 1 3/4 2-5 Amendment No.61

POWER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 F,"(X,Y,Z) shall-be evaluated to determine if F,(X,Y,Z) is within its limit by:

a.

Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER; b.

Measuring F,"(X,Y,Z) at the earliest of:

1.

At least once per 31 Effective Full Power Days, or 2.

After exceeding by 20% or more of RATED THERMAL POWER the THERMAL POWER at which F,"(X,Y,Z) was last determined *;

c.

Satisfying the relationship presented in Specification 3.2.2; d.

Satisfying the following relationship:

F,"(X,Y,Z) $ [F,(X,Y,Z)]""

where [F (X,Y,Z)]"* represents the nominal design power distribution increase 0 by an allowance for the expected deviation between the nominal design power distribution and the measurement and is specified in the COLR.

If the above relationship is not satisfied, then for that location perform the following:

1.

Calculate the % margin to the maximum allowable design as i

follows:

F "(X,Y,Z)

% Opera ti onal Marg i n - ( 1 - --------,=

= - --

)

100

[F,'(X, Y, Z)]*

F,"(X, Y, Z)

% Reactor Protection - ( 1 - ----

t

)

100 Setpoint (RPS)

[ F,'(X, Y, Z ) ]""'

Margin where, [F,'(X,Y,Z)]" and [F,'(X,Y,2)]ars are the Operational and RPS design peaking limits and are specified in the COLR.

2.

Find the minimum Operational Margin of all locations examined in 4.2.2.2.d.1, above.

If the minimum margin is less than 0, EITHER of the following actions shall be taken:

During power escalation at the beginning of each cycle, THERMAL POWER may be increased until a power level for extended operation has been achieved and a power distribution may be obtained.

WOLF CREEK - UNIT 1 3/4 2-6 Amendment No. 61

L POWER DISTRIBUTION LIMITS SURVEILLANCE RE0VIREMENTS (Continued) a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , control the AFD to within new AFD limits that are determined by:

Reduced negative AFD Limit -

The negative AFD Limit in Specification 3.2.1 plus the absolute value of the quantity

[Op Mar NSLOPE

Minimum Operational Margin],

Reduced positive AFD Limit -

The positive AfD Limit in Specification 3.2.1 minus the absolute value of the quantity

[Op Mar PSLOPE

Minimum Operational Margin],

where, the Op Mar NSLOPE and 0p Mar PSLOPE are specified in the COLR, and declare the AFD monitor alarm inoperable until the AFD alarm setpoints are modified to the limits of 4.2.2.2.d.2.a or b.

Comply with the ACTION requirements of Specification 3.2.2, treating the margin violation in 4.2.2.2.d.1, above, as the amount by which F,"(X,Y,Z) is exceeding its limit.

3.

Find the minimum RPS margin of all locations examined in 4.2.2.2.d.1, above.

If the minimum margin is less than 0,-the.

following action shall be taken:

Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , reduce the negative f (AI) limit and the 3

positive f (AI) limit of the OTAT as follows:

3 Reduced negative f (AI) Limit -

3 f (AI) of Table 2.2-1 3

plus the absolute value of the quantity

[the RPS Mar NSLOPE

Minimum RPS Margin],

Reduced positive f (AI) Limit -

3 f (AI) of Table 2.2-1 3

minus the absolute value of the quantity

[the RPS Mar PSLOPE

Minimum RPS Margin],

WOLF CREEK - UNIT 1 3/4 2-7 Amendment No.61

i

+

i POWER DISTRIBUTION LIMITS

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SURVEILLANCE RE0VIREMENTS (Continued)

{

where, RPS Mar NSLOPE and RPS Mar.PSLOPE-are specified in the COLR.

e.

The limits in Specification 4.2.2.2.d are not applicatie in the j

following core plane regions as measured in percent of core height r

from the bottom of the fuel:

!?

q 1.

Lower core region from 0 to 15%,. inclusive, 2.

Upper core region from 85 to 100%, inclusive, 3.

Grid Plane Regions, and 4.

Core plane regions within +/- 2% of core height;(+/ 2.88 inches) about the bank demand position of the Bank "D" control.

rods.

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WOLF CREEK - UNIT 1 3/4 2-8 Amendment No. 61-

F POWER DISTRIBUTION LIMITS l

3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - F,n(X,Y)

LIMITING CONDITION FOR OPERATION t

3.2.3 F,,(X,Y) shall be limited by the following relationship:

FMR"(X,Y) s FMR'(X,Y)

where, i

FMR"(X,Y) = the maximum measured radial peak ratio defined in the Core Operating Limits Report (COLR).

FMR'(X,Y) = the maximum allowable radial peak ratio defined and specified in the COLR.

APPLICABILITY: MODE 1 ACTION:

With F,,(X,Y) exceeding its limit:

a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , reduce the allowable THERMAL POWER from RATED THERMAL POWER at least RRH%* for each 1% that FMR"(X,Y) exceeds the limit, and b.

Within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either:

1.

Restore FMR"(X,Y) to within the limit for RATED THERMAL POWER, or-2.

Reduce the Power Range Neutron Flux - High Trip Setpoint at least RRH% for each 1% that FMR"(X,Y) exceeds that limit, and c.

Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of initially being outside the limit, either:

1.

Restore FMR"(X,Y) to within the limit for RATED THERMAL POWER, or 2.

Perform the following actions:

a.

Reduce the OTAT K, term by at least TRH** for each 1% that FMR"(X,Y) exceeds the limit, and b.

Verify through incore mapping that FMR"(X,Y) is restored to within the limit for the THERMAL POWER allowed by ACTION a, 3

or reduce THERMAL POWEP, to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and

RRH is the amount of THERMAL POWER reduction required to compensate for each 1% that FMR"(X,Y) exceeds FMR'(X,Y) and is specified in the COLR.

- TRH is the amount of OTAT K, ds the limit and is specified in the COLR.setpoint r each 1% that FMR"(X,Y) excee WOLF CREEK - UNIT 1 3/4 2-9 Amendment No. 23, 61

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued) d.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a and/or c.2.b, above; subsequent POWER OPERATION may proceed provided that EoHR"(X,Y) is demonstrated, through incore flux mapping, to be within the limit specified in the COLR prior to exceeding the following THERHAL POWER levels:

1.

A nominal 50% of RATED THERMAL POWER, 2.

A nominal 75% of RATED THERHAL POWER, and 3.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal to 95% of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 E6HR"(X,Y) shall be evaluated to determine whether F,(X,Y) is within 3

its limit by:

Measuring E6HR"(X,Y) according to the following schedule:

a.

1.

Prior to operation above 75% of RATED THERMAL POWER at the beginning of e:ch cycle, and 2.

At least once per 31 Effective Full Power Days.

b.

Satisfying the following relationship:

FMR"(X,Y) s; FMR" "(X,Y) where, E6HR"*(X,Y) represents the nominal design power distribution increased by an allowance for the expected deviation between the nominal design power distribution and the measurement and is specified in the COLR.

l If the above relationship is not satisfied, then for that location I

perform the following.

1.

Calculate the % margin to the maximum allowable design as follows:

FMR"(X,Y)

% F,, Marg i n - ( 1 - --------------

-) 100 FoHR'(X,Y) 2.

Find the minimum margin for all locations examined in 4.2.3.2.b.1, above.

If the minimum margin is less than 0, comply with the ACTION requirements of Specification 3.2.3.

WOLF CREEK - UNIT 1 3/4 2-10 Amendment No. 61 i

4 POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION i

ACTION (Continued) i 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High i

Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and 3.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is v'erified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified at 95% or greater RATED THERMAL POWER.

d.

The provisions of Specification 3.0.4 are not applicable.

SUR'IEILLANCE-REQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:

a.

Calculating the ratio at least once'per 7 days when the alarm is OPERABLE, and b.

Calculating the ratio at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during steady-state operation when the alars is inoperable.

4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range Channel inoperable by using the movable incore detectors to confirm that the normalized symmetric power distribution, obtained from two sets of four symmetric thimble locations or a full-core flux map, is consistent with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

WOLF CREEK - UNIT 1 3/4 2-13

F POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1:

Reactor Coolant System T,,,

a.

b.

Pressurizer Pressure, and c.

Reactor Coolant System (RCS) Flow Rate APPLICABILITY: MODE 1.*

i ACTION:

i r

a.

With parameter 1 or 2 of Table 3.2-1 exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

With the RCS total flow rate outside the region of acceptable operation shown on Table 3.2-1:

1.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

a.

Restore the total flow rate to within the above limit, or b.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER.

2.

Within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either:

a.

Restore the total flow rate to within the above limit, or b.

Reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER.

1 3.

Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of initially being outside the above limit, verify that the RCS total flow rate is restored to within the above limit, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next-2 hours; and i

See Special Test Exception Specification 3.10.4 for 3.2.5.c.

WOLF CREEK - UNIT 1 3/4 2-14 Amendment No. 61

1 POWER DISTRIBUTION LIMITS l

r 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPER*, TION ACTION:

(Continued) 4.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION 1.b and/or 3, above; subsequent POWER OPERATION may proceed provided that the indicated RCS total flow rate is demonstrated to be within the region of acceptabie operation prior to exceeding the following THERMAL POWER levels:

a.

A nominal 50% of RATED THERMAL POWER, i

b.

A nominal 75% of RATED THERMAL POWER, and L

c.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal to 95%

of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.2.5.1 The provisions of Specification 4.0.4 are not applicable to Specification 3.2.5.c.

4.2.5.2 Each of the parameters of Table 3.2-1 shall be verified to be within their limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.2.5.3 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.

4.2.5.4 The RCS total flow rate shall be determined by precision heat balance measurement at least once per 18 months. Within 7 days prior to performing the precision heat balance, the instrumentation used for determination of steam pressure, feedwater pressure, feedwater temperature, and feedwater venturi AP in the calorimetric calculations shall be calibrated.

4.2.5.5 The feedwater venturi shall be inspected for fouling and cleaned as necessary at least once per 18 months.

t i

WOLF CREEK - UNIT 1 3/4 2-15 Amendment No. 61 s

O TABLE 3.2-1 DNB PARAMETERS LIMITS j

Four Loops in-PARAMETER Operation 1.

Indicated Reactor Coolant System T,,

5592.5*F 2.

Indicated Pressurizer Pressure 22220 psig*

3.

Reacter Coolant System Flow Rate 238.4 x 10' GPM

+

?

Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.

WOLF CREEK - UNIT 1 3/4 2-16 Amendment No. 61.

(

INSTRUMENTATION MOVABLE INCORE DETECTORS LIMITING CONDITION FOR OPERATION 3.3.3.2 The Movable Incore Detection System shall be OPERABLE with:

a.

At least 75% of the detector thimbles, i

b.

A minimum of two detector thimbles per core quadrant, and c.

Sufficient movable detectors, drive, and readout equipent to map these thimbles, j

b APPLICABILITY: When the Movable Incore Detection System is used for:

l l

a.

Recalibration of the Excore Neutron Flux Detection System, b.

Monitoring the QUADRANT POWER TILT RATIO, or 1

Measurement of F,(X,Y,Z) and F,,(X,Y).

c.

ACTION:

a.

With the Movable Incore Detection System inoperable, do not use the system for the above applicable monitoring or calibration functions, b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS 4.3.3.2 The Movable Incore Detection System shall be demonstrated OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by normalizing each detector output when required for:

a.

Recalibration of the Excore Neutron Flux Detection System, or b.

Monitoring the QUADRANT POWER TILT RATIO, or Measurement of F,(X,Y,Z) and F,(X,Y).

c.

4 WOLF CREEK - UNIT 1 3/4 3-43 Amendment No.61

)

INSTRUMENTATION I

SEISMIC INSTRUMENTATION I

f LIMITING CONDITION FOR OPERATION 3.3.3.3 The seismic monitoring instrumentation shown in Table 3.3-7 shall be OPERABLE.

APPLICABILITY: At all. times.

ACTION:

With one or more of the above required seismic monitoring instruments a.

inoperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the instrument (s) to OPERABLE status, b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable, i

SURVEILLANCE REQUIREMENTS 4.3.3.3.1 Each of the above required seismic monitoring instruments shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-4.

4.3.3.3.2 Each of the above required seismic monitoring instruments. actuated during a seismic event greater than or equal to 0.01 g shall be restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and a CHANNEL CALIBRATION performed within t

10 days following the seismic event.

Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground motion.

A Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 14 days describing the magnitude, fre-quency spectrum, and resultant effect upon facility features important to safety.

[

WOLF CREEK - UNIT 1 3/4 3-44

L 1

TABLE 3.7-2 STEAM LINE SAFETY VALVES PER LOOP I

i VALVE NUMBER LIFT SETTING * (f3%)** ORIFICE SIZE l

l Looo 1 Loon 2 Looo_3 Looo 4 I

V055 V065 V075 V045 1185 psig 16.0 sq. in.

V056 V066 V076 V046 1197 psig 16.0 sq. in.

l V057 V067 V077 V047 1210 psig 16.0 sq. in.

V058 V068 V078 V048 1222 psig 16.0 sq. in.

f V059 V069 V079 V049 1234 psig 16.0 sq. in.

I r

I i

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

- After testing, the valves will be left at il%.

WOLF CREEK - UNIT 1 3/4 7-3 Amendment No. 61 l

}

PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION t

3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

Two motor-driven auxiliary feedwater pumps, each capable of being a.

powered from separate emergency busses, and t

b.

One steam turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.

l APPLICABILITY:

MODES 1, 2, and 3.

ACTION:

a.

With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With two auxiliary feedwater pumps inoperable, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With three auxiliary feedwater pumps inoperable, immediately initiate c.

corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible.

SURVEILLANCE REQUIREMENTS 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:

a.

At least once per 31 days on a STAT.iERED TEST BASIS by:

1)

Verifying that each motor-driven pump develops a discharge pressure of greater than or equal to 1535 psig on recirculation flow when tested pursuant to Specification 4.0.5; 2)

Verifying thst the steam turbine-driven pump develops a discharge pressure of greater than or equal to 1625 psig at a flow of greater than or equal to 120 gpm when the secondary steam supply pressure is greater than 900 psig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3; i

WOLF CREEK - UNIT 1 3/4 7-4

L 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION i

LIMITING CONDITION FOR'0PERATION 3.9.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met:

a.

A K,,, of 0.95 or less, or b.

A boron concentration of greater than or equal to the limit specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY: MODE 6*.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal-to 7000 ppm boron or its equiv-alent until K, is reduced to less than or equal to 0.95 or the boron concen-tration is re,s,tored to greater than or equal to the limit specified in the COLR, whichever is the more restrictive.

SURVEILLANCE RE0UIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:

a.

Removing or unbolting the reactor vessel head, and b.

Withdrawal of any full-length control rod in excess of 3 feet from its fully inserted position within the reactor vessel.

4.9.1.2 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

4.9.1.3 Valves BG-V178 and BG-V601 shall be verified locked closed and secured in position at least once per 31 days.

The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with 3

the head removed.

WOLF CREEK - UNIT 1 3/4 9-1 Amendment No. 61

s f

f REFUELING OPERATIONS l

3/4.9.2 INSTRUMENTATION-LIMITING CONDITION FOR OPERATION i

3.9.2 As a minimum, two source range neutron flux monitors shall be OPERABLE r

each with continuous visual indication in the control room and one with audible i

indication in the containment and control room.

APPLICABILITY:

MODE 6.

ACTION:

a.

With one of the above required monitors inoperable or not operating, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

b.

With both of the aFove required monitors inoperable or not operating, determine the boron concentration of the Reactor Coolant System at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

t SURVEILLANCE REQUIREMENTS 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of:

l a.

A CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , b.

An ANALOG CHANNEL OPERATIONAL TEST within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to the initial start of CORE ALTERATIONS, and c.

An ANALOG CHANNEL OPERATIONAL TEST at least once per 7 days.

~

a 6-3 WOLF CREEK - UNIT 1 3/4 9-2 i

i

t REFUELING OPERATIONS 3/4.9.12 SPENT FUEL ASSEMBLY STORAGE LIMITING CONDITION FOR OPERATION 3.9.12 Spent fuel assemblies stored in Region 2 shall be subject to the following conditions:

a.

The combination of initial enrichment and cumulative exposure shall be within the acceptable domain of Figure 3.9-1, and b.

No spent fuel assemblies shall be placed in Region 2, nor shall any storage location be changed in designation from being in Region I to being in Region 2, while refueling operations are in progress.

APPLICABILITY:

Whenever irradiated fuel assemblies are in the spent fuel pool.

ACTION:

With the requirements of the above specification not satisfied, a.

suspend all other movement of fuel assemblies and crane operations with loads in the fuel storage areas and move the non-complying fuel assemblies to Region 1.

Until these requirements of the above specification are satisfied boron concentration of the spent fuel pool shall be verified to be greater than or equal to 2000 ppe at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.12 The burnup of each spent fuel assembly stored in Region 2 shall be ascertained by analysis of its burnup history and independently verified, prior to storage in Region 2.

A complete record of such analysis shall be kept for the time period that the spent fuel assembly remains in Region 2 of the spent fuel pool.

WOLF CREEK - UNIT 1 3/4 9-15

50 i

r

~

40 B

30 R

[

=au.

b UnacceptaNe

20 10 Erychment Burnup 2.1 10.572 2.6 18.069

~

3.1 25.14 3.8 34.G01 4.5 43.357 L

0 1.5 2

2.5 3

3.5 4

4.5 5

ENRICHMENT (w/o)

FIGURE 3.9-1 WOLF CREEK MINIMUM REQUIRED FUEL ASSEMBLY BURNUP AS A FUNCTION OF INITIAL ENRICHMENT FOR STORAGE IN REGION 2 WOLF CREEK - UNIT 1 3/4 9-16 Amendment No. IE, 61

r f

y.

i i

3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and SHUTDOWN MARGIN provided reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod (s).

APPLICABILITY:

MODE 2.

ACTION:

a.

With any full-length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

b.

With all full-length control rods fully inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or its equivalent until the SHUTDOWN HARGIN required by Specification 3.1.1.1 is restored.

SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full-length control rod either partially or fully withdrawn shall be determined at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

4.10.1.2 Each full-length control rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1.

i WOLF CREEK - UNIT 1 3/4 10-1

i i

SPECIAL TEST EXCEPTIONS 3/4.10.2 GROUP HEIGHT. INSERTION. AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION I

3.10.2 The group height, insertion, and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:

a.

The THERMAL POWER is maintained less than or equal to 85% of RATED THERMAL POWER, and b.

The limits of Specifications 3.2.2, 3.2.3, and 3.2.5.c are maintained and determined at the frequencies specified in Specification 4.10.2.2, below.

r APPLICABILITY: MODE 1.

ACTION:

With any of the limits of Specifications 3.2.2, ' 2.3, or 3.2.5.c being l'

exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 are suspended, either:

a.

Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2, 3.2.3, and 3.2.5.c, or l

b.

Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE0VIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

4.10.2.2 The requirements of the below listed specifications siall be performed at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during PHYSICS TESTS-a.

Specification 4.2.2.2, b.

Specification 4.2.3.2, and c.

Specification 4.2.5.2.

)

r WOLF CREEK - UNIT 1 3/4 10-2 Amendment No. 61 t

i

L SPECIAL TEST' EXCEPTIONS 3/4.10.3. PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specifications 3.1.1.3, 3.1.1.4, 3.1.3.1, 3.1.3.5, and 3.1.3.6 zay be suspended during the performance of PHYSICS TESTS provided:

The THEhtKAL POWER does not exceed 5% of RATED THERMAL

,a.

b.

The Reactor Trip Setpoints on the OPERABLE Intersediate and Power Range channels are set at less than or equal to 25% of RATED THERMAL POWER, and The Reactor Coolant System lowest operating loop temperature (T c.

is greater than or equal to 541*F.

APPLICAEILITY: MODE 2.

ACTION:

With the THERMAL POWER greater than 5% of RATED THERHAL POWER, a.

inmediately open the Reactor trip breakers, b.

With a Reactor Coolant System operating loop temperature (T,yg) les than 541'F, restore T,yg to within its limit within 15 minutes or be in at least HDT STANDBY within the next 15 minutes.

SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERFAL POWER shall be determined to be less than or equal to 5%

of RATED THEPyAL POWER at least once per hour during PHYSICS TESTS.

4.10.3.2 Each Intermediate and Power Range channel shall be subjected to an ANALOG CHANHEL OPERATIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating PHYSICS TESTS.

4.10.3.3 The Reactor Coolant Systen temparature (Tug) shall be deterstned to be greater than or equal to 541'F at least once per 30 minutes during PHYSICS TESTS.

W3LF CREEK - UNIT 1 3/4 10-3

j r

SPECIAL TEST EXCEPTIONS 3/4.10.4 REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION I

3.10.4 The limitations of the following requirements may be suspended:

a Specification 3.2.3, 3.2.5.c and 3.4.1.1 - During the performance of startup and PHYSICS TESTS in MODE 1 or 2 provided:

1)

The THERMAL POWER does not exceed the P-10 Interlock Setpoint, and 2)

The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are set less than er equal to 25% of RATED THERMAL POWER.

b.

cecification 3.4.1.2 - During the performance of hot rod drop time measurements in MODE 3 provided at least three reactor coolant loops as listed in Specification 3.4.1.2 are OPERABLE.

APPLICABILITY: During operation below the P-10 Interlock Setpoint or performance of hot rod drop time measurements.

ACTION:

a.

With the THERMAL POWER greater than the P-10 Interlock Setpoint during the performance of startup and PHYSICS TESTS, immediately open the Reactor trip breakers.

~

3 b.

Vith less than the above required reactor coolant loops OPERABLE during performance of hot rod drop time measurements, immediately place two reactor coolant loops in operation.

SURVEILLANCE REQUIREMENTS 4.10.4.1 The THERMAL POWER shall be determined to be less than P-10 Interlock Setpoint at least once per hour during startup and PHYSICS TESTS._

4.10.4.2 Each Intermediate and Power Range channel, and P-10 Interlock shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating startup and PHYSICS TESTS.

4.10.4.3 At least the above required reactor coolant loops shall be determined OPERABLE within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior. to initiation of the hot rod drop time measurements and at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during the hot rod drop time measurements by verifying correct breaker alignments and indicated power availability.

WOLF CREEK - UNIT 1 3/4 10-4 Amendment No. S7, 35,61 i

e

b 3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 B0 RATION CCNTROL 3/4.1.1.1 and 3!4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that: (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients asso-ciated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T condition occurs at EOL, with T,, at no load opera *t'in.

The most restrictive g temperature, and is associated with a postulated steam line break accident and resulting uncon-trolled RCS cooldown.

In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.3% ok/k is required to control the reactivity transient. Accord-ingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With T.

less than 200*F, the reactivity transients resulting from a postulated stea,m line break cooldown are minimal and a 1.3% ok/k SHUTDOWN MARGIN provides adequate i

protection.

I 3/A.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident and transient _ analyses.

The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.

s WOLF CREEK - UNIT 1 B 3/4 1-1 Amendment No. 61

REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued)

The most negative MTC value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions. These corrections involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions. This value of the MDC was then transformed into the limiting MTC End of Life (E0L) value specified in the CORE OPERATING LIMITS REPORT (COLR). The 300 ppm surveillance limit MTC value specified in the COLR represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting E0L MTC value specified in the COLR.

The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since tivis coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 551*F.

This limitation is required to ensure: (1) the moderator temperature coefficient is within it analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel is above its i

minimum RT temperature.

gg 3/4.1.2 B0 RATION SYSTEMS The Boration Systems ensures that negative reactivity control is available during each mode of facility operation. The components required to

+

perform this function include: (1) borated water sources, (2) centrifugal charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature equal to or greater than 350'F a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoper-able. The boration capability of either flow path is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.3% ok/k after xenon decay and cooldown to 200*F. The maximum expected boration capability require-ment occurs at E0L from full power equilibrium xenon conditions and requires 17,658 gallons of 7000 ppm borated water from the boric acid storage tanks or 83,754 gallons of 2400 ppm borated water from the RWST. With the RCS average temperature less than 350', only one boron injection flow path is required.

WOLF CREEK - UNIT 1 B 3/4 1-2 Amendment No. 23, 61

i REACTIVITY CONTROL SYSTEMS BASES B0 RATION SYSTEMS (Continued)

With the RCS temperature below 200*F, one Boration System is acceptable without single failure consideration on the basis of the stable reactivity con-dition of the reactor and the additional restrictions prohibiting CORE ALTERA-TIONS and positive reactivity changes in the event the single Boron Injection System becomes inoperable.

The limitation for a ma3 ' mum of one centrifugal charging pump to be OPERABLE and the Surveillance P.equirement to verify all charging pumps except the required OPERABLE pump to be inoperable in MODES 4, 5, and 6 provides assur-ante that a mr.ss addition pressure transient can be relieved by the operation of a single PORV or an RHR suction relief valve.

The boron capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN of 1.3% ok/k after xenon decay and cooldown from 200*F to 140*F.

l' This condition requires either 2968 gallons of 7000 ppm borated water from the boric acid storage tanks or 14,071 gallons of 2400 ppm borated water from the RWST.

The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

In the.

case of the boric acid tanks, all of the contained volume is considered usable.

The required usable volume may be contained in either or both of the boric acid tanks.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The OPERABILITY of one Boration System during REFUELING ensures that this system is available for reactivity control while in MODE 6.

3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is main-tained, and (3) the potential effects of rod misalignment on associated accice 't analyses are limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with

- 1 the control rod alignment and insertion limits.

Verification that the Digital Rod Position Indicator. agrees with the demanded position within i 12 steps at 24, 48, 120, and 228 steps withdrawn for the Control Banks and 18, 210 and 228 steps withdrawn for the Shutdown Banks provides assurances that the Digital Rod Position Indicator is operating correctly over tLe full' range of indication.

Since the Digital Rod Position System does not indicate the actual shutdown rod position between 18 steps and 210 steps, only points in the indicated ranges are picked for verification of agreement with demanded position.

WOLF CREEK - UNIT 1 B 3/4 1-3 Amendment No. 23, 61

l REACTIVITY CONTROL SYSTEMS BASES i

MOVABLE CONTROL ASSEMBLIES (Continued)

For purposes of determining compliance with Specification 3.1.3.1, any immovability of a control rod invokes ACTION Statement 3.1.3.1.a.

Before utilizing ACTION Statement 3.1.3.1.c, the rod control urgent failure alars must be illuminated or an electrical problem must be detected in the rod control system.

The rod is considered trippable if the rod was demonstrated OPERABLE during the last performance of Surveillance Requirement 4.1.3.1.2 and met the rod drop time criteria of Specification 3.1.3.4 during the last performance of Surveillance Requirement 4.1.3.4.

The ACTION statements which permit limited variations from the basic reqcirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires measurement of peaking factors and a restriction in THERMAL POWER. These restrictions provide assurance of fuel rod integrity during continued operation.

In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation.

The power reduction and shutdown time limits given in ACTION statements 3.1.3.2.a.2, 3.1.3.2.b.2, and 3.1.3.2.c.2, respectively, are initiated at the tirre of discovery that the compensatory actions required for POWER OPERATION can no longer be met.

The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Measurement with T,yg greater than or equal to 551*F and with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions.

I Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with more frequent verificatio'ns required if an automatic monitoring channel is inoperable.

These verification frequencies te adequate for assuring that the applicable LCOs are satisfied.

t WOLF CREEK - UNIT 1 B 3/4 1-4 Amendment No. 27,46

i 3/4.2 POWER DISTRIBUTION LIMITS I

BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:

(a) maintaining the minimum DNBR in the core greater than or equal t

to the DNBR design limit specified in the CORE OPERATING LIMITS REPORT (COLR) i during normal operation and in short-term transients, and (b) limiting the fission gas release, fuel pellet temperature, and cladding mechanical proper-ties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

I F,(X, Y, Z)

Heat Flux Hot Channel Factor, is defined as the local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods, at assembly (X,Y);

F,,(X, Y)

Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power at assembly (X,Y).

3/4.2.1 AXIAL FLUX DIFFERENCE i

The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F (X,Y,Z) and F,,(X,Y) limits are not exceeded during either normal operation o,r in the event of xenon redistribution following power changes. The AFD limits have l

been adjusted for measurement uncertainty.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer deter-i mines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the AFD limits and the THERMAL POWER is greater tnan 50% of RATED THERMAL POWER.

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor and nuclear enthalpy rise hot channel factor ensure that: (1) the design limits on peak local power density and minimum DNBR are not exceeded, and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criteria limit.

WOLF CREEK - UNIT 1 B 3/4 2-1 Amendment No. J, 61 i

t POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

Each of these is measurable but will normally only be determined

[

periodically as specified in Specifications 4.2.2 and 4.2.3.

This periodic surveillance is sufficient to insure that the limits are maintained provided:

a.

Control rods in a single group move together with no individual rod insertion differing by more than 12 steps, indicated, from the group demand position, b.

Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6, c.

The control rod insertion limits of Specification 3.1.3.6 are maintained, and d.

The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

F (X,Y) will be maintained within its limits provided Conditions a.

a through d. above are maintained. The limits on the nuclear enthalpy rise hot channel factor, F X,Y), are specified in the COLR as Maximum Allowable Radial Peak Ratio l(imits, obtained by dividing the Maximum Allowable Peak u

(MAP) limit by the axial peak for assembly location (X,Y).

By definition, the Maximum Allowable Radial Peak Ratio limits will result in a DNBR for the limiting transient that is equivalent to the DNBR calculated with the design F (X,Y) value specified in the COLR and a limiting reference axial power u

shape.

F"XYZ that the(y,re,ma)in within their limits.and FAHR"(X,Y) are measured periodically to A peaking margin calculation is performed, when necessary, to provide the basis for reducing THERMAL POWER, for reducing the width of the AFD limits, and for reducing the f (AI) limits 3

of the OTAT trip setpoints.

3/4.2.4 OUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribution satisfies the design values used in the power capability analysis.

Radial power distribution measurements are made during STARTUP testing and periodically during power operation.

The limit of 1.02, at which corrective ACTION is required, provides DNB and linear heat generation rate protection with x-y plane power tilts. A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.

WOLF CREEK - UNIT 1 B 3/4 2-2 Amendment No. I, EJ. 61

t i

POWFR DISTRIBUTION LIMITS BASES OVADRANT POWER TILT RATIO (Continued)

The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod.

In the event such ACTION does not. correct the tilt, the margin for uncertainty on F,(X,Y,Z) is reinstated by reducing the maximum allowed power by 3% for each percent of tilt in excess of 1.

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that

-the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8.

3/4.2.5 DNB PARAMETERS The limits on the Reactor Coolant System T, and the pressurizer pressure assure that each of the parameters are, maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial USAR assumptions and have been analytically demonstrated adequate to maintain a CNBR above the safety analysis limit DNBR specified in the CORE OPERATING LIMITS REPORT (COLR) throughout each analyzed transient. The indicated T value of 592.5*F and the indicated pressurizer pressure value of 222E,psig correspond to analytical limits of 595'F and 2205 psig respectively, with allowance for measurement uncertainty.

The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

Fuel rod bowing reduces the value of DNB ratio. Credit is available to offset this reduction in the generic margin. The generic margins completely offset any rod bow penalties. This is the margin between the correlation DNBR limit and the safety analysis limit DNBR. These limits are specified in the COLR.

The applicable values of rod bow penalties are referenced in the USAR.

When RCS flow rate and F (X,Y), per Specification 3.2.3, are measured, a

no additional allowances are necessary prior to comparison with the limits in the COLR. Measurement uncertainties of 2.5% for RCS total flow rate and 4%

for F,,(X,Y) have been allowed for in determination of the design DNBR value.

WOLF CREEK - UNIT I B 3/4 2-3 Amendment No. EI,61 e

POWER DISTRIBUTION LIMITS BASES DNB PARAMETERS (Continued)

The measurement uncertainty for RCS total flow rate is based upon performing a precision heat balance and using the result-to calibrate the RCS flow rate indicators.

Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a nonconservative manner. Therefore, an inspection is performed of the feedwater venturi each refueling outage.

The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the acceptable region of operation specified in Table 3.2-1.

This surveillance also provides adequate monitoring to detect any core crud buildup.

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WOLF CREEK - UNIT 1 8 3/4 2-4 Amendment No. 61

n INSTRUMENTATION BASES REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)

If they are, the signals are combined into logic metrices sensitive to combina-tions indicative of various accidents, events, and transients. Once the re-quired logic combination is completed, the system sends actuation signals to i

those Engineered Safety Features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss-of-coolant accident: (1) Safety Injection pumps start and automatic valves position, (2) Reactor trip, (3) Feed-water System isolates, (4) the emergency diesel generators start, (5) contain-ment spray pumps start and automatic valves position, (6) containment isolates, (7) steam line. isolation, (8) Turbine trip, (9) auxiliary feedwater pumps start and automatic valves position, (10) containment cooling fans start and auto-matic valves position, (11) essential service water pumps start and automatic valves position, and (12) isolate normal control room ventilation and start Emergency Ventilation System.

Encineered Safety Features Actuation System Interlocks i

The Engineered Safety Features Actuation System interlocks perform the following functions:

P-4 Reactor tripped - Actuates Turbine trip, closes main feedwater valves on T,yg below Setpoint, prevents the opening of the main feedwater valves which were closed by a Safety Injection or High Steam Generator Water Level signal, allows Safety Injection block so i

that components can be reset or tripped.

Reactor not tripped prevents manual block of Safety Injection.

P-11 On increasing pressure P-11 automatically reinstates safety injection actuation on low pressurizer pressure and low steamline pressure and automatically blocks steamline isolation on negative steamline pressure rate.

On decreasing pressure; P-11 allows the manual block of Safety Injection on low pressurizer pressure and low steamline pressure and allows steamline isolation on negative steamline pressure rate to become active upon manual block of low steamline pressure SI.

WOLF CREEK - UNIT 1 B 3/4 3-3 Amendment No. 43 l

1

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INSTRUMENTATION BASES l

3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING FOR PLANT OPERATIONS The OPERABILITY of the radiation monitoring instrumentation for plant operations ensures that: -(1) the associated ACTION.will be initiated when the radiation level monitored by each channel-or combination thereof reaches.its Setpoint, (2) the specified coincidence-logic is maintained, and -(3) sufficient redundancy is maintained to permit a channel to be out-of-service.

l for testing or maintenance. The radiation monitors for plant operations senses radiation levels in selected plant systems and locations and determines whether or not predetermined limits are being exceeded.

If they are, the

~!

signals are combined into logic matrices sensitive to combinations indicative of various accidents and abnormal conditions..Once the required logic combination is completed, the system sends ~ actuation signals to. initiate alarms or automatic isolation action and actuation of Emergency Exhaust or Control Room Emergency Ventilation Systems.

3/4.3.3.2 MOVABLE INCORE DETECTORS

+

The OPERABILITY of the movable incore detectors with the specifie'd minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.

j For the purpose of measuring F "(X,Y,Z) or'F "(X,Y) a full incore flux

-l o

p map is used. - Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the Excore Neutron Flux Detection System,_and full i

incore' flux maps or symmetric incore thimbles may be used for monitoring the-QUADRANT POWER TILT RATIO when one Power Range Neutron Flux channel is inoperable.

l 3/4.3.3.3 SEISMIC INSTRUMENTATION

.The OPEP1.BILITY of the seismic instrumentation ensures that sufficient' i

capability:is available to promptly determine the magnitude of a seismic event j

and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that-

-used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix A of 10 CFR Part 100. The instrumentation-is j

consistent with the recommendations of Regulatory Guide 1.12, " Instrumentation -

.l for Earthquakes," April 1974.

i i

WOLF CREEK - UNIT 1 B 3/4 3-4 Amendmint No. 61 j

l J

l 4

DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The core shall contain 193 fuel assemblies.with each fuel-assembly normally containing 264 fuel rods clad with Zircaloy-4 except-that limited substitution of fuel rods by filler rods consisting of Zircaloy-4 or stainless L

steel or by vacancies may be made if justified by a cycle-specific reload analysis.

Each fuel rod shall have a nominal active fuel length of 144 inches. The initial core loading shall have a maximum nominal enrichment of i

3.10 weight percent U-235.

Reload fuel shall be similar in physical design to i

the initial core loading.

I CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 53 full-length and no part-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material. All control rod assemblies shall be hafnium, silver-indium-cadmium, or a mixture of both types. All control rods shall be clad with stainless steel tubing.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:

a.

In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the 1

applicable Surveillance Requirements, b.

For a pressure of 2485 psig, and c.

For a temperature of 650*F, except for the pressurizer which is t

680*F.

C VOLUME 5.4.2 The total volume of the Reactor Coolant System, including pressurizer and surge line, is 12,135 100 cubic feet at a nominal T,, of 557'F.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

j F

WOLF CREEK - UNIT 1 5-6 Amendment No. 3. IE,19, 48,61

l DESIGN FEATURES l

5.6 FUEL STORAGE CRITICALITY 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with:

I a.

A k,,, equivalent to less than or equal to 0.95 when flooded with unborated water, which includes an allowance for uncertainties as described in Section 4.3 of the USAR. This is based on new fuel with an enrichment of 4.45 weight percent U-235 in Region I and on spent fuel with combination of initial enrichment and discharge exposures, shown in Figure 3.9-1, in Region 2, and b.

A nominal 9.236 inch center-to-center distance between fuel assemblies placed in the storage racks.

5.6.1.2 The k,,, for new fuel for the first core loading stored dry in the j

spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed.

DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 2040 feet.

l CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1344 fuel assemblies.

l s

5.7-COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be l

maintained within the cyclic or transient limits of Table 5.7-1.

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WOLF CREEK - UNIT 1 5-7 Amendment No. 16,61

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THIS PAGE HAS BEEN DELETED i

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F WOLF CREEK - UNIT I 5-8 Amendment No.16,61 r

I ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (COLR) 6.9.1.9 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT (COLR) before each reload cycle or any remaining part of a reload cycle, for the following:

1.

Specification 3.1.1.3: Moderator Temperature Coefficient (MTC) E0L limits 2.

Specification 3.1.3.5:

Shutdown Rod Insertion Limit 3.

Specification 3.1.3.6: Control Rod Insertion Limits 4.

Specification 3.2.1:

Axial Flux Difference (AFD) 5.

Specification 3.2.2:

Heat Flux Hot Channel Factor - F,(X,Y,Z) 6.

Specification 3.2.3:

Nuclear Enthalpy Rise Hot Channel Factor -

F (X,Y) u 7.

Specification 3.9.1.b:

Refueling Boron Concentration

[

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents.

a.

NRC Safety Evaluation Report dated October 29, 1992, for the " Core Thermal Hydraulic Analysis Methodology for the Wolf Creek Generating Station" (ET-90-0140, ET 92-0103)

(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor - F (X,Y) u b.

NRC Safety Evaluation Report (upon issuance) for the " Transient Analysis Methodology for the Wolf Creek Generating Station" (ET 0026, ET 92-0142, WM 93-0010, WM 93-0028)

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient [MTC))

c.

NRC Safety Evaluation Report dated March 26, 1993, for the

" Qualification of the Steady State Core Physics Methodology for the l

Wolf Creek Generating Station" (ET 92-0011, WM 93-0038)

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient (MTC); Specification 3.1.3.5 - Shutdown Rod Insertion Limit; Specification 3.1.3.6 - Control Rod Insertion Limits; Specification 3.2.1 - Axial Flux Difference; Specification 3.2.2 -

Heat Flux Hot Channel Factor - F (X,Y,Z); Specification 3.2.3 -

o Nuclear Enthalpy Rise Hot Channel Factor - F (X,Y); Specification u

3.9.1.b - Refueling Boron Concentration)

P WOLF CREEK - UNIT 1 6-21 Amendment No. 42, 61-i

~

ADMINISTRATIVE CONTROLS 2

CORE OPERATING LIMITS REPORT (COLR) (Continued) d.

NRC Safety Evaluation Report dated March 10, 1993, for the " Reload Safety Evaluation Methodology for the Wolf Creek Generating Station" (ET 92-0032, ET 93-0017)

(Methodology for Specification 3.1.3.6 - Control Rod Insertion Limits; Specification 3.2.1 - Axial Flux Difference) e.

NRC Safety Evaluation Report dated March 30, 1993, for the " Revision to Technical Specification for Cycle 7" (NA 92-0073, NA 93-0013, NA 93-0054)

(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor - F (X,Y), [Use of WRB-2 Correlation with VIPRE-01 u

Code])

f.

NRC Safety Evaluation Report dated November 13, 1986, for "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code" (WCAP-10266-P-A, Rev. 2)

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel I

Factor - F,(X,Y,Z)

The core operating limits shall be determined so that all applicable I

limits (e.g., fuel thermal-hydraulic limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, j

to the NRC Document Control Desk with copies to the Regional Administrator and Resident inspector.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report.

6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

1 6.10.1 The following records shall be retained for at least 5 years:

j a.

Records and logs of unit operation covering time interval at each

)

power level; i

l b.

Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety; WOLF CREEK - UNIT 1 6-21a Amendment No. 61 j

i

t ADMINISTRATIVE CONTROLS RECORD RETENTION (Continued) li c.

All REPORTABLE EVENTS; d.

Records of surveillance activities, inspections, and calibrations required by these Technical Specifications; e.

Records of changes made to the procedures required by Specification 6.8.1; f.

Records of radioactive shipments; g.

Records of sealed source and fission detector leak tests and results; and h.

Records of annual physical inventory of all sealed source material i

of record.

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I i

WOLF CREEK - UNIT 1 6-21b Amendment No.61

ADMINISTRATIVE CONTROLS RECORD RETENTION (Continued) 6.10.2 The following records shall be retained for the duration of the Unit Operating License:

Records and drawing changes reflecting unit design modifications a.

made to systems and equipment described in the Final Safety Analysis Report; b.

Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories; Records of radiation exposure for all individuals entering radiation c.

control areas; d.

Records of gaseous and liquid radioactive material released to the environs; Records of transient or operational cycles for those Unit components e.

identified in Table 5.7-1; f.

Records of reactor tests and experiments; g.

Records of training and qualification for current members of the Unit Staff; h.

Records of in service inspections performed pursuant to these Technical l

Specifications; i.

Records of Quality Assurance activities required by the QA Manual; j.

Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59; k.

Records of meetings of the PSRC and the NSRC; 1.

Records of the service lives of all hydraulic and mechanical snubbers required by Specification 3.7.8 including the date at which the service life commences and associated installation and maintenance records; Records of secondary water sampling and water quality; and m.

Records of analysis required by the Radiological Environmental n.

Monitoring Program that would permit evaluation of the accuracy of the analysis at a later date.

This should include procedures effective at specified times and QA records showing that these procedures were followed, Records of reviews performed for changes made to the OFFSITE DOSE i

o.

CALCULATION KANUAL amt the PROCESS CONTROL PROGRAM.

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{

WOLF CREEK UNIT 1 6-22 Amendment No. 42 l

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ML20035C619

Text

Attachment:

0.5 Where

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